Volume 82, Issue 5

The European Human Resource Observatory for Nuclear (EHRO-N) provides the European Commission with essential data related to supply and demand for nuclear experts in the EU-28 and the enlargement and integration countries based on bottom-up information from the nuclear industry. The objective is to assess how the supply of experts for the nuclear industry responds to the needs for the same experts for present and future nuclear projects in the region. Complementary to the bottom-up approach taken by the EHRO-N team at JRC, a top-down modelling approach has been taken in a collaboration with NRG in the Netherlands. This top-down modelling approach focuses on the human resource requirements for operation, construction, decommissioning, and efforts for long term operation of nuclear power plants. This paper describes the top-down methodology, the model input, the main assumptions, and the results of the analyses.

The best-estimate KARATE code system has been widely used for core design calculations and simulations of slow transients of VVER reactors. Recently there has been an increasing need for assessing the uncertainties of such calculations by propagating the basic input uncertainties of the models through the full calculation chain. In order to determine the uncertainties of quantities of interest during the burnup, the statistical version of the KARATE code system has been elaborated. In the first part of the paper, the main features of the new code system are discussed. The applied statistical method is based on Monte-Carlo sampling of the considered input data taking into account mainly the covariance matrices of the cross sections and/or the technological uncertainties. In the second part of the paper, only the uncertainties of cross sections are considered and an equilibrium cycle related to a VVER-440 type reactor is investigated. The burnup dependence of the uncertainties of some safety related parameters (e.g. critical boron concentration, rod worth, feedback coefficients, assembly-wise radial power and burnup distribution) are discussed and compared to the recently used limits.

The Fukushima Daiichi accident occurring on March 11, 2011, reveals that Station Blackout (SBO) may last longer than 8 h. However, the original design may not have sufficient capacity to cope with a SBO for more than 8 h. In view of this, Taiwan Power Company has initiated several enhancements to mitigate the severity of the extended SBO. Based on the improved plant configuration, a SBO coping analysis is performed in this study to assess whether the Kuosheng BWR plant has sufficient capability to cope with SBO for 24 h with respect to maintaining the integrity of the reactor core and containment. The analyses in the Nuclear Steam Supply System (NSSS) and the containment are based on the RETRAN-3D and GOTHIC models, respectively. The flow conditions calculated by RETRAN-3D during the event are retrieved and input to the GOTHIC containment model to determine the containment pressure and temperature response. These boundary conditions include SRV flow rate, SRV flow enthalpy, and total reactor coolant system leakage flow rate.

Turbulent mixing rate between adjacent subchannels in a two-phase flow has been known to be strongly dependent on the flow pattern. The most important aspect of turbulent motion is that the velocity and pressure at a fixed point do not remain constant with time even in steady state but go through very irregular high frequency fluctuations. These fluctuations influence the diffusion of scalar and vector quantities. The Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type, heavy water moderated and boiling light water cooled natural circulation based reactor. The fuel bundle of AHWR contains 54 fuel rods set in three concentric rings of 12, 18 and 24 fuel rods. This fuel bundle is divided into number of imaginary interacting flow channel called subchannels. Alteration from single phase to two phase flow situation occurs in reactor rod bundle with raise in power. The two phase flow regimes like bubbly, slug-churn, and annular flow are generally encountered in reactor rod bundle. Prediction of thermal margin of the reactor has necessitated the investigation of turbulent mixing rate of coolant between these subchannels under these flow regimes. Thus, it is fundamental to estimate the effect of spacer grids on turbulent mixing between subchannels of AHWR rod bundle.

In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1 X -A1, U 3 O 8 -Al, and U 3 Si 2 -Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1 X -A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U 3 Si 2 -Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

The Fukushima nuclear disaster in 2011 has raised widespread concern over the safety of nuclear power plants. This study employed knowledge management in conjunction with the Teoriya Resheniya Izobreatatelskih Zadatch (TRIZ) method in the formulation of a database to facilitate the evaluation of post-fire safe shutdown capability with the aim of safeguarding nuclear facilities in the event of fire. The proposed approach is meant to bring facilities in line with US Nuclear Regulatory Commission (NRC) standards. When implemented in a case study of an Asian nuclear power plant, our method proved highly effective in the detection of 22 cables that fell short of regulatory requirements, thereby reducing 850,000 paths to 0. This study could serve as reference for industry and academia in the development of systematic approaches to the upgrading of nuclear power plants.

In the case of Fukushima Daiichi nuclear power plant (FNP) accident, the radioactive material was released from reactor units 1–3 and transported to short and long distances due to the atmospheric pathways-motions. Power sources for monitoring posts were lost due to earthquake and tsunami. Based on air dose rates and other data measured by monitoring cars, the amount of radioactive material released to the atmosphere from the power station was obtained. The atmospheric dispersion and the transport model used in the RASCAL code, estimate the radionuclide concentrations downwind, both in the air and on the ground due to deposition. The calculated concentrations are then used to estimate the projected doses for workers in vicinity of the accident area in the first minutes of accident time. For dose modeling, we assumed that each worker was 15 min in vicinity of FNP in accident situation, once without and once with protective clothes or respirator. According to Tokyo Electric Power Company (TEPCO) report six workers had received doses over 250 mSv (309 to 678 mSv) apparently due to inhaling Iodine-131 fume. In this paper the calculated dose results using RASCAL code shows that, if emergency workers who work in early phase of accident had not used protective equipment, for 15 min, inhalation doses from iodine in their thyroid gland up to 12 March afternoon would have been 520 mSv. A comparison between calculation results and TEPCO report shows that dose calculated virtually is nearly equal to TEPCO measurement results.

Owing to growing interest in the study of emitted antineutrinos from nuclear reactors to test the Atomic Energy Agency safeguards, antineutrino flux was studied in the Tehran Research Reactor (TRR) using ORIGEN code. According to our prediction, antineutrino rate was obtained 2.6 × 10 17 (ν¯e${\bar \nu _e}$/sec) in the core No. 57F of the TRR. Calculations indicated that evolution of antineutrino flux was very slow with time and the performed refueling had not an observable effect on antineutrino flux curve for a 5 MW reactor with the conventional refueling program. It is seen that for non-proliferation applications the measurement of the contribution of 239 Pu to the fission using an antineutrino detector is not viable in the TRR.

The friction force between the contact surfaces of a reactor internal hold-down spring (HDS) and core barrel flanges can directly influence the axial stiffness of an HDS. However, friction coefficient cannot be obtained through theoretical analysis. This study performs a mathematical deduction of the physical model of an HDS. Moreover, a mathematical model of axial load P , displacement δ , and flexibility coefficient is established, and a set of test apparatuses is designed to simulate the preloading process of the HDS. According to the experimental research and theoretical analysis, P -δ curves and the flexibility coefficient λ are obtained in the loading processes of the HDS. The friction coefficient f of the M1000 HDS is further calculated as 0.224. The displacement limit load value (4,638 kN) can be obtained through a displacement limit experiment. With the friction coefficient considered, the theoretical load is 4,271 kN, which is relatively close to the experimental result. Thus, the friction coefficient exerts an influence on the displacement limit load P . The friction coefficient should be considered in the design analysis for HDS.

In nuclear reactor imbalance of axial power distribution induces xenon oscillations. These fluctuations must be maintained bounded within allowable limits. Otherwise, the nuclear power plant could become unstable. Therefore, bounded these oscillations is considered to be a restriction for the load following operation. Also, in order to design the nuclear reactor control systems, poisons concentrations, especially xenon must be accessible. But, physical measurement of these parameters is impossible. In this paper, for the first time, in order to estimate the axial xenon oscillations and ensures these oscillations are kept bounded within allowable limits during load-following operation, a robust observer based nonlinear control based on multipoint kinetics reactor model for pressurized-water nuclear reactors is presented. The reactor core is simulated based on the multi-point nuclear reactor model (neutronic and thermal-hydraulic). Simulation results are presented to demonstrate the effectiveness of the proposed observer based controller for the load-following operation.

The objective of this paper is to estimating radionuclide concentrations in air and radiological dose consequences to indoor workers in a containment of MTR open pool type reactor during emergency situations. A postulated core degradation accident causes fission products to release and the ventilation system will be converted automatically from the normal situation into emergency situation in order to purify the contaminated air by forcing it to pass through a group of filters. The study computes internal and external worker doses from inhalation and submersion in a finite cloud of contaminated air during the emergency mode. A radiological toolbox version 2 was used to evaluate the radiation dose levels inside the containment.

One speed, time independent and homogeneous medium neutron transport equation is solved with the anisotropic scattering which includes both the linearly and the quadratically anisotropic scattering kernel. Having written Case's eigenfunctions and the orthogonality relations among of these eigenfunctions, slab albedo problem is investigated as numerically by using Modified F N method. Selected numerical results are pesented in tables.